Shielding materials for fusion reactors

ABSTRACT

There is described neutron shielding for a nuclear fusion reactor. The neutron shielding includes a cemented carbide or boride comprising a binder and an aggregate, the aggregate comprising particles of a carbide or boride compound.

CROSS-REFERENCE TO RELATED APPLICATIONS

The present application is a divisional application of U.S. patentapplication Ser. No. 15/315,211, filed on Nov. 30, 2016, the entirecontents of which are incorporated herein by reference and priority towhich is hereby claimed. Application Ser. No. 15/315,211 is the U.S.National Stage of Application No. PCT/GB2015/051961, filed Jul. 7, 2015.Priority under 35 U.S.C. § 119(a) and 35 U.S.C. § 365(b) is herebyclaimed from United Kingdom Application No. 1412540.5, filed Jul. 15,2014 and United Kingdom Application No. 1505156.8, filed Mar. 26, 2015,the disclosures of which are both also incorporated herein by reference.

TECHNICAL FIELD

The present invention relates to neutron shielding materials for fusionreactors. In particular, though not exclusively, the invention relatesto shielding for use in a compact spherical tokamak reactor.

BACKGROUND

The challenge of producing fusion power is hugely complex. Fusionneutrons are produced when a deuterium-tritium (D-T) ordeuterium-deuterium (D-D) plasma becomes very hot so that the nucleifuse together, releasing energetic neutrons. To date, the most promisingway of achieving this is to use a tokamak; in the conventional tokamakapproach to fusion (as embodied by ITER), the plasma needs to have highconfinement time, high temperature, and high density to optimise thisprocess.

A tokamak features a combination of strong toroidal magnetic field BT,high plasma current I_(p) and usually a large plasma volume andsignificant auxiliary heating, to provide a hot stable plasma so thatfusion can occur. The auxiliary heating (for example via tens ofmegawatts of neutral beam injection of high energy H, D or T) isnecessary to increase the temperature to the sufficiently high valuesrequired for nuclear fusion to occur, and/or to maintain the plasmacurrent.

The problem is that, because of the large size, large magnetic fields,and high plasma currents generally required, build costs and runningcosts are high and the engineering has to be robust to cope with thelarge stored energies present, both in the magnet systems and in theplasma, which has a habit of ‘disrupting’—mega-ampere currents reducingto zero in a few thousandths of a second in a violent instability.

The situation can be improved by contracting the donut-shaped torus of aconventional tokamak to its limit, having the appearance of a coredapple—the ‘spherical’ tokamak (ST). The first realisation of thisconcept in the START tokamak at Culham demonstrated a huge increase inefficiency—the magnetic field required to contain a hot plasma can bereduced by a factor of 10. In addition, plasma stability is improved,and build costs reduced.

WO 2013/030554 describes a compact spherical tokamak for use as aneutron source or energy source. An important consideration in thedesign of spherical tokamaks is the protection of reactor componentsfrom the high neutron flux generated by the fusion reaction. This is ofparticular importance on small tokamaks as the neutron flux (i.e.neutron flow per unit area) will in general be higher due to the smallersurface area-to-volume ratio of the plasma vessel.

The present application is based on a very compact form of the tokamak,and employs a range of innovative features, including use of HighTemperature

Superconducting magnets. The ‘Efficient Compact Fusion Reactor’ (ECFR)is intended to provide a compact fusion power plant. FIG. 1 is aschematic diagram of such a reactor. The plasma (11) is contained withina vacuum vessel (12) by the magnetic fields generated by a toroidalfield coil (13) and a poloidal field coil (not shown). The toroidalfield coil runs down a central column (14) in the centre of the plasmachamber.

A drawback of the ST is that the limited space in the central columnprohibits installation of the substantial shielding necessary to protectthe central windings in a neutron environment—so conventional toroidalfield windings, and conventional central solenoids (used to induce andmaintain the plasma currents) are not practical. Although power plantsbased on the ST have been designed (using solid copper centre posts withlimited shielding, the post being changed every year or so when damagedby neutrons), these have high energy dissipation in the centre columndue to the relatively high resistivity of warm copper, requiring a largedevice for electricity production to become economical.

Superconducting materials may be used for the central core, but suchmaterials are vulnerable to damage from neutrons, and may failcatastrophically if enough damage accumulates that the material nolonger superconducts. There is therefore a trade-off between the overallsize of the central core, the cross sectional area of thesuperconducting material (which is related to the maximum current thatthe superconductor can carry), and the thickness of the shielding.

In order to ensure that the reactor is as compact as possible (whichallows greater efficiency), the thickness of shielding should be reducedas much as possible, while still maintaining adequate protection for theother components. Minimising the distance between the plasma and thefield coils allows a higher magnetic field in the plasma with a lowercurrent in the coils.

FIG. 2 shows a section of the central column, and illustrates theproblems which the shielding material must overcome. The central column(13) comprises a central core of HTS coils (21) and an outer layer ofshielding (22). Depending on the material used for the shielding, theremay be a layer of oxidised shielding material (23) on the outer surface.There are three major causes of damage which originate from the plasma.Firstly, the high energy neutrons generated by the fusion reaction canessentially knock atoms out of the structure of the shielding, creatingdamage cascades which propagate through the material. Secondly, the heatflux from the fusion reaction is significant, and can damage theshielding due to thermal stresses induced by uneven heating and the HTScore, as higher temperatures reduces the current that can be carriedwhile maintaining superconductivity, and can cause the coil to suddenlygain resistance, causing the magnet to quench. Lastly, the energeticparticles of the plasma will ablate the outer surface of the shielding.This not only causes damage to the shielding itself, but can alsocontaminate the plasma. It is desirable to have a shielding materialwhich can resist these effects, as well as prevent neutrons fromreaching the superconducting coils.

SUMMARY

The challenge of providing shielding in a compact spherical tokamak isaddressed in this document by proposing alternative materials which maybe used for neutron shielding. The focus will be on shielding for thecentral column, as this is the most space-critical region of thespherical tokamak, but it is anticipated that the disclosure can beeasily adapted for use with other components of the reactor.

According to a first aspect, there is provided neutron shielding for anuclear fusion reactor, the neutron shielding including a cementedcarbide or boride comprising a binder and an aggregate, the aggregatecomprising particles of a carbide or boride compound of tungsten,tantalum or hafnium.

According to a further aspect, there is provided a central column for anuclear fusion reactor, the central column comprising neutron shieldingaccording to the first aspect and a core of superconducting material,wherein the neutron shielding is arranged to protect the superconductingmaterial from heating and damage by neutrons.

According to a still further aspect, there is provided a compact nuclearfusion reactor. The reactor comprises a toroidal plasma chamber and aplasma confinement system arranged to generate a magnetic field forconfining a plasma in the plasma chamber. The plasma confinement systemis configured so that the major radius of the confined plasma is 1.5 mor less and an aspect ratio of the plasma is 2.5 or less. The neutronshielding used to prevent neutron damage to sensitive components of thereactor is neutron shielding according to the first aspect.

According to a still further aspect, there is provided a divertor for anuclear fusion reactor, the divertor including a cemented carbide orboride comprising a binder and an aggregate, the aggregate comprisingparticles of a carbide or boride compound of tungsten, tantalum orhafnium.

According to a still further aspect, there is provided the use of acemented carbide or boride as neutron shielding for a nuclear fusionreactor, the cemented carbide or boride comprising a binder and anaggregate, the aggregate comprising particles of a carbide or boridecompound of tungsten, tantalum or hafnium.

Further aspects and preferred features are set out in the appendedclaims.

BRIEF DESCRIPTION OF THE DRAWINGS

FIG. 1 is a schematic diagram of a spherical tokamak reactor

FIG. 2 is a schematic diagram of a section of reactor shielding.

DETAILED DESCRIPTION

In order to be suitable as for use as shielding in a fusion reactor, amaterial should be good at absorbing fusion-energy neutrons, resistantto thermal shock, resistant to sputtering and plasma ablation, andresistant to neutron damage. Two classes of materials, the use of whichis proposed in this document, which would appear to have all of theseproperties, are cemented carbides and cemented borides.

Cemented carbides are a metal matrix composite in which particles of acarbide act as the aggregate, and a metallic binder serves as thematrix. Cemented carbides are formed by a sintering process, in whichthe material is heated to a point where the binder is liquid, but thecarbide particles remain solid. The carbide grains are thereby embeddedinto the liquid binder, which is then allowed to set. This results in amaterial with superior qualities to either the carbide or the bindertaken alone. The ductile binder offsets the natural brittleness of thecarbide ceramic, and the carbide particles make the resulting compositemuch harder than the binder alone. Due to the metal binder, cementedcarbides typically have a high thermal conductivity, which reduces thethermal stress experienced by the material due to uneven heating. Thecoefficient of linear thermal expansion of cemented carbides or boridesis typically in the range of 4 to 5×10⁻⁶. Cemented materials are alsoresistant to sputtering (ablation of the outer surface of the materialby energetic particles). For example, cemented tungsten carbidetypically has one quarter of the sputtering rate of pure tungsten.

Cemented borides are equivalent, but using boride particles as theaggregate, rather than carbide. Borocarbide particles may also be used.

The choice of carbide/boride and binder will be guided by the conditionsin the reactor. The need to withstand high neutron flux prevents the useof many elements and isotopes, such as cobalt and nickel, which wouldbecome radioactive due to neutron exposure. High magnetic fields requirestructural considerations to be taken into account when usingferromagnetic material, as the resulting forces would cause largestresses within the reactor. Similar considerations occur for the choiceof carbide. Also, the material must of course be able to reduce the fluxof neutrons which reach components behind the shield. Carbon willnaturally act as a moderator, slowing the fission neutrons down, whichallows greater freedom of choice in the other elements that may be used(since many more elements are effective absorbers of slow neutrons thanfaster neutrons). Boron-10 is an effective neutron absorber.

Promising candidates for the carbide are tungsten carbide, as theneutron absorption is favourable and the mechanical properties have beenwell studied, tungsten boride, and boron carbide, which combines themoderating properties of carbon with the neutron absorption of boron.Multiple carbides may be used in order to balance structural andneutronics properties of the material. In addition, other substances maybe added to the cemented material in addition to the carbides, forexample borides may be added to a predominantly carbide composite inorder to introduce boron into the shielding, or vice versa. Addition oftungsten boride to a cemented tungsten carbide may improve theresistance to corrosion. Borocarbides which may be used include tungstenborocarbide, specifically a ternary tungsten borocarbide. Othersubstances that may be added to the material include oxides andnitrides, for example titanium nitride may be added to improve thestructural properties of the material.

Other alternatives to tungsten carbide or tungsten borocarbide includeborides and/or carbides of elements corresponding to the third long rowof the periodic table (or beyond). The melting points of the elementsincrease across the third period, peaking at group six (tungsten).Therefore the main candidate elements are hafnium, tantalum, tungstenand rhenium. The platinum metals may be theoretically suitable forneutron shielding but are considered to be less useful because osmiumcompounds are highly toxic, and because of the prohibitively high costof iridium and platinum. Rhenium is also very expensive and very rare.The three most likely candidates are therefore hafnium, tantalum andtungsten. Of these, tungsten (including its compounds) is the cheapestand most widely available, and easy to process by powder methods.

Tantalum has better ductility and toughness than tungsten, is easier toform and join (e.g. by welding), and has better oxidation resistance.However, it is a scarce material and very expensive to buy, and becomesmuch more radioactive than tungsten when exposed to fusion energyneutron irradiation. The activity decays to below tungsten levels afterabout a hundred years, but that is an unacceptably long time. Hafnium isalso useful. Hafnium diboride is highly refractory, and has very goodoxidation resistance. Hafnium is quite rare, but can be obtained as aside product of the production of zirconium for the nuclear industry.

In a spherical tokamak it is important to use a shielding material richin tungsten (or other element from the third long row of the periodictable, such as hafnium or tantalum) because of the space limitations.This, in turn, creates an acute problem in terms of oxidation andcorrosion resistance (because tungsten oxidation is exothermic, and theoxide is volatile). The incorporation of borides (and/or silicides) intoa tungsten or tungsten carbide based shield helps address this problem.

The composition of the shielding may be graded, for example the outer(i.e. plasma facing) regions of the shielding may be formulated toimprove the resistance to corrosion and ablation, whereas the innerregions may be formulated to improve structural properties or thermaltransport. This may be used to improve the efficiency of the shielding,for example by including a higher concentration of neutron moderatingmaterial towards the outside (i.e. plasma facing side) of the shielding,and a higher concentration of neutron absorbing material towards theinside of the shielding. In this way, the neutron absorbers are placedwhere the neutrons will be slowest, and the absorbers will be mosteffective. Such finely graded structures would be difficult if notimpossible to achieve with conventional alloying techniques, and providea further advantage to the use of cemented materials.

The manufacturing process for cemented carbides or borides allowscomplex structures to be made relatively easily compared to manufacturefrom other materials. For example, it would be simple to build shieldingwith holes through which coolant could be run. Furthermore, cementedcarbides or borides may be joined to other materials by a variety oftechniques, including brazing and specialised welding methods (e.g.electron beam or laser welding). This provides a considerable advantagewhen manufacturing the overall reactor systems, e.g. to join theshielding to the main structure of the reactor.

Some aspects of the use of cemented carbides/borides may seemcounter-intuitive, but careful study reveals that these aspects do notin fact pose a problem. For example, the metals used for the binder (apromising combination is iron and chromium) have a relatively lowmelting point compared to other materials used in construction of thereactor, and it is not inconceivable that parts of the shielding will beraise above the melting point. However, if the binder melts, the carbideparticles will tend to hold it together until it re-freezes in situ.Even in the extreme case where the binder volatilises on the plasmafacing side, the carbide will form a solid shell, which will maintainthe structure of the shielding (though the thermal performance may beimpacted on the outer layers).

Furthermore, it may seem that the use of powdered carbide/boride wouldnot produce a uniform enough substance for the neutronics to befavourable. However, provided the mean free path of the neutrons issubstantially greater than the diameter of any individual particle inthe cemented material then the powder blend will act identically to a“true” alloy. The mean free path of the neutrons is one or two orders ofmagnitude greater than the particle sizes which are used for cementedcarbides.

Cemented carbides or borides may also be used in other regions of thereactor, e.g. the divertor, where they provide similar advantages.

Although the invention has been described in terms of preferredembodiments as set forth above, it should be understood that theseembodiments are illustrative only and that the claims are not limited tothose embodiments. Those skilled in the art will be able to makemodifications and alternatives in view of the disclosure which arecontemplated as falling within the scope of the appended claims. Eachfeature disclosed or illustrated in the present specification may beincorporated in the invention, whether alone or in any appropriatecombination with any other feature disclosed or illustrated herein.

1. A divertor for a nuclear fusion reactor, the divertor including a cemented carbide or boride comprising a binder and an aggregate, the aggregate comprising particles of a carbide or boride compound of tungsten, tantalum or hafnium.
 2. A divertor according to claim 1, wherein the composition of the cemented carbide or boride varies through the thickness of the cemented carbide or boride.
 3. A divertor according to claim 2, wherein a plasma-facing side of the cemented carbide or boride comprises a lower proportion of neutron absorbing material than a non-plasma facing side of the cemented carbide or boride.
 4. A divertor according to claim 1, wherein the binder comprises a metal.
 5. Neutron shielding according to claim 4, wherein the binder comprises iron and/or chromium.
 6. Neutron shielding according to claim 1, wherein the cemented carbide or boride does not comprise cobalt or nickel. 